ML19221A663
| ML19221A663 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/21/1979 |
| From: | Anderson T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Ross D Office of Nuclear Reactor Regulation |
| References | |
| NS-TMA-2088, NUDOCS 7905250393 | |
| Download: ML19221A663 (21) | |
Text
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Westinghouse Water Reactor w 5:'m m ~ s :n Electric Corporation Divisions g, m P "mrt1F :r.na um May 21, 1979 NS-TMA-20e8 Mr. Denwood F. Ross, Jr.
Deputy Director Division of Project Management Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555
Subject:
ACRS Recommendations Relating to TMI-2 Accident
Dear Mr. Ross:
Enclosed you will find the Westinghouse responses to the Advisory Com-mittee on Reactor Safeguards (ACRS) recommendations relating to the TMI-2 accident. These responses were requested by your letter dated May 17, 1979.
If you have any questions regarding the enclosed, don't hesitate to contact this office.
Very truly yours, f
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- %w T. M.'Ande'rson, Manager Nuclear Safety Department
/bek Enclosure 0 001se e Anack ed
- 1 7 905250 3D
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A.
Letter, M. Carbon to Chairman Hendrie, dated April 7, 1979 Recommendation 1 - Perform further analyses of small break transients and accidents.
Westinghouse Response The response to this recommendation will be divided into two separate parts:
1) small break analyses performed by W
- 2) TMI Scenario
- 1) W has perf ormed, prior to TMI incident, numerous small break analyses, botn pient specific and generic, for various conditions.
To illustrate the renge of these small break analyses, Table 1 lists some of those analyses.
This list is not intended to be all inclu-sive of the number of small breaks analyses performed.
The previously reported analyses for breaks in the pressurizer vapor space indicated pressurizer level holdup for breaks equivalent to relief valve size and associated pipe size.
These analyses indi-cated that a substantial period of time without safety injection can be tolerated before core uncovery is predicted to occur.
Therefore, the operator has sufficient time to manually initiate safety injec-tion.
Subsequent to TMI additional s.nall breaks have been performed to verify previous calculations with latest models, respond to NRC request and to provide our customers with generic information.
Typical of the pressurizer vapor space analyses performed are sum-marized in Table 2 for an equivalent 2 1/2" diameter break using Appendix K assumptions.
These results reconfirm the behavior of the W system to small pres-surizer breaks and demonstrate the amount of time available for safety injection initiation prior to any core uncovery.
166 234
In conclusion, besides typical SAR small break analyses, an addi-tional spectrum of breaks have been performed to determine system response and acceptability
- 2) Calculations have been performed on a W system using a TMI type transient. The following scenario and assumptions were used:
Step 1:
- Initiating event is loss of feedwater flow (0-2000 seconds) - Control systems assumed available
- Realistic physics assumed
- Trip on steam generator lo-level coincident with steam / feed flow mismatch
- Assume no auxiliary feedwater flow Results:
- S/G heat transfer begins to degrade to the point where system heats up (<1200 sec)
- PCRV calculated to cpen at #2000 sec (If aux.
feedflow was available, P03V would not open.)
- Liquid still present in SG Secondary Step 2:
(>2000 seconds)
- Using the results of Step 1 as initial conditic.is, the PORV is assumed to stick open
- Assume loss of offsite power
- Auxiliary feedwater and safety injection initiated
- Appendix K assumptions Results:
- Slight core uncovery 0
- PCT remains significantly belcw 1S00 F
- System reaches stable condition with core covered.
166 2315
TABLE 1 EXM!PLES OF SMALL BREAK ANALYSES PERFORMED TOPIC /DATE BREAK SIZE BREAK LOCATION (Equivalent Diameter)
PESAR 3/1972 3",
4",
6" Cold Leg 1",
3",
6" Pressurizer Vapor Space TVA - MIP/197 4 3",
4",
6" Hot Leg WCAP-8356/1974 2",
3",
4',
Cold Leg 6",
8",
0.5 ft 1 ft for 2,
3 and 4 loop plants 8" - 4 loop Hot Leg 6" - 3 loup Hot Leg 4" - 2 loop Hot Leg RESAR 414/1978 0.5",
3",
4" Cold Leg 6",
8" WCAP-8970/1979 2",
3",
4",
6" Cold Leg for 2, 3 and 4 loop plants 166 236 e
TABLE 2 PRESSURIZER VAPOR SPACE BREAKS
- 2 1/2" Diameter Hole
- Appendix K Assumptions PLANT TYPE SI ACTUATED
SUMMARY
OF RESULTS 4 loop Yes
- No core uncovery No
- Top of core uncovers e2200 sec 2 loop Yes
- No core uncovery No
- Top of core uncovers
- 1800 sec 3 loop Yes
- No core uncovery No
- Estimated top of core uncovers e 2000 sec i66 237
Recommendation 2 - Provide operator additional informaticn and means to follow the course of an accident; as a minimum, consider expeditiously:
(a) unambiguous RV level indication (b) remotely controlled vent for RCS high points 2(a) Westinghouse Response Westinghouse is evaluating the feasibility and desirability of imple-mentation of a reactor vessel water level indicator.
This evaluation is not complete, however, the following comments are of note to date:
The availability of sach information could be ceneficial to an operator, should a situation occur in which the several levels of protection in the plant have been violated.
Tnis indication would have to be qualifiec for all conditions of plant operation including forced and natural circulation ficw and single phase and two-hase flow.
As a result, the instrument design would be complex.
Installa ion of such instrumentation would be performed in a high radiation area. Maintenance and in-service testing, cali-bration, etc. would also be conducted in a high radiation area.
As a result, increases in occupational radiation expo-sures (ORE) would occur.
These ORE increases should be con-sidered as part of an evaluation of the cost / benefit of imple-mentation of such an instrument.
Westinghouse has reviewed the present standard designs of engi-neered safety features anc has concluded that initiation ano control of the safety systems can all be accomplished under the Westinghouse recommended Emergency Operating Procedures utiliz-ing existing instrumentation.
166 238
2(b) Westinghouse Response The Westinghouse design philosophy for the Reactor Coolant System ano safety systems is to prevent occurrence of a situation in which large volumes of non-condensible gases would occur.
In adoition, the Reactor Coolant System layout criteria result in a reactor vessel location such that the top of the core is approximately S feet below the reactor vessel hot leg / cold leg nozzle centerline.
The Reactor Coolant System hot legs are horizontal piping runs connected to the irlet of the U-tube steam generators.
The pressurizer surge line is connected to the horizontal centerline, or above the centerline, of one hot leg and is required to be continuously sloped upward to the pressurizer.
Effectively, the reactor vessel head, hot leg piping, pressurizer, prassurizer surge line and steam generators are all located above tne Reactor Coolant System centerline.
As a result, should it ceccme necessary to remove gas from the system, it is possible to perform the degassing through the pressurizer without uncovering the reactor core.
The high points in the f? actor Coolant System are the Reactor Vessel Head, the Pressurizer and the Steam Generator U-Dends.
In the West nghouse system, the Pressurizer is the highest point, ano can ce vented through the Power Operated Relief Valves.
The remaining high point in the system for which venting would be practical is the reactor vessel head.
Westinghouse is in the process of evaluating the desirability of installing a remote venting system in the reactor vessel head.
Although this cvaluation is not complete, Westinghouse does offer the following comments:
Installation of such a system, if properly designed, would provide an increase in post accident cperation flexibility.
i66 239
Design and installation of such a system would provide an appro-priate means for satisfying portions of Regulatory Guice 1.139 oy providing a remote safety grace letdown capaaility.
The system could ccmplicate refueling operations and in-service inspection requirements, and could lead to increases in Occupa-tional Radiation Exposures for these activities.
The layout for such a syste.a would have to consider seismic pro-tection, pipe whip, 39t impingement and other dynamic effects.
Installation on h oack-fit basis would represent significant impacts on operating personnel exposures, routing of equipment, shielding design, etc.
166 240
Recommendation 3 -
Item 4b of Bulletin 79-05A considered unduly prescriptive in '!iew of uncertainties in predicting course of anomalous small break transients / accidents.
Westinghout Response Bulletin 79-05A does not apply to plants with Westinghouse NSSS.
How-ever, Westinghouse has prepared recommended modifications for Emergency Operating Procedures to provide for termination of safety injection where continued operation could potentially lead to unsafe plant condi-tions.
For additional details on these recommended modifications, see Westinghouse letter NS-TMA-2087, May 17, 1979, T. M. Anderson to D. F. Ross.
166 241
B.
Letter, R. Fraley to Commissioners, dated April 18, 1979 Recommenda: ion 1 - Natural Circulation-related Items a.
Detailed analyses of natural circulation mode, supported as required oy experiment, by licen-sees and NSSS vendors.
b.
Develop procedures for initiating natural cir-culation.
c.
Provide operator means for assurance that natu-ral circulation has been established, e.g.,
by installation of instructions to indicate flow at low velocities.
d.
Expeditiously survey operating PWR's to deter-mine wnether suitable arrangements of PZR heaters and reliable on-site power distribution can be provided to assure this vital aspect of natural circulation capability.
e.
Operator should be adequately informed concern-ing RCS conditions which affect natural circula-tion capability, e.g.,
(1) indication that RCS is approaching satura-tion condition by suitable display to oper-ator of T nd T and PZR pressure in c
b conjunction with steam tables (2) use of flow exit temperature indicator fuel assembly thermocouples, where available.
i66 242
1(a) Westinghouse Response Westinghouse has developed computer models of natural circulation in 2, 3, and 4 loop plants.
The results of these calculations have been veri-fied by plant test data.
The basic principle used in natural circulation cooling is that the change in density due to convective cooling / heating as coolant flows through the core and steam generators results in a ouoyancy driving head (Figure 1).
Therefore design mu. assure that when in this cooling mode resultant density gradients occur at the most advantageous equipment elevations (Steam Generators elevated aoove reactor core).
Westinghouse has always recognized the high importance of natural circulaticn folloa-ing loss of pumping power to assure removal of core decay neat without violation of NRC criteria.
Therefore layout and equipment design criteria hai/e always been set consistent with enhancing this mode of cooling.
Natural circulation is enhanced by ensuring a large height difference (AZ) between heat generation and h - loss and minimizing component flow resistance.
The models used to predict natural circulation for safety analyses and equipment layout input calculations have also required extensive review and verification programs.
This review has included direct submittal of the models to the NRC as part of the computer code approval for safety analysis calculations and the Westinghouse application of these flow models to specific accident conditions.
In addition, the ability of these models to conservatively predict natural circulation flows has been independently verified and documentea at several operating plants.
These tests have been performed en: 2 loop 14x14, 3 loop ISxis, nna 4 loop 15x15 plants (Figure 2).
166 20
FIGURE 1 STEADY STATE NATURAL CIRCULATION Flow is defined as:
-2 1/3 2g D 0 0 03 C
W=
C 2 (K/
23) p With:
AZ - Height between heat generation and heat loss Q
- Decay heat generated p
- Average density B
- Volumetric coefficient of thermal expansion K
- Con 1ponent flow resistance A
- Component flow area 166 244
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1 1(b) L (c) Westinghouse Response Natural circulatica has been shown to occur in Westinghouse "WR's by the performance of tests and during blackout and other abnormal conditions that have occurrec at operating plants. Westinghouse agrees that natu-ral circulation guidelines should be added to existing procedures that deal with abnorma, operating conditions in which natural circulation may become the core heat removal mode. A discussion of proposed guidelines follows.
Natural circulation is enhanced when primary water inventory is main-tained and the reactor coolant is subcooled.
To accomplish this it is re ommended that pressurizer level be increasing or g 50% and pressur-izer pressure be > 2000 psia.
A pressure of 2000 psia. results in at least 15 subcooling at the core outlet at a 100% power T value hat 0
(e 620 ).
With the liquid subcooled, void due to steam or noncon-densible gas formation will not be present so with indicated level in the pressurizer we are certain the active reactor coolant system is water solid.
For natural circulation to occur a heat sink must be present.
By main-taining level in the same steam generators at a point above the top of the tubes, i.e. in the narrow range, secondary side heat sink is also maintained.
There is indication available to the operator that will allow him to determine if natural circulation is occurring.
Since the natural cir-culation flow to power ratio exceeds that during normal operation, RCS AT with natural circulation in progress will ce < full pcwer AT.
If natural circulation were lost the AT would increase as Tnct increased. Another indication of heat removal via natural circulation is constant or decreasing primary temperatures as read on reactor cool-ant T indicators or on core exit thermocouples.
If natural circu-hot lat.on were lost core outlet temperatures >;ould increase.
One other indication of maintenance of natural circulation would be steam pressure remaining constant or decreasing at the same rate as prirury tempera-tures while maintaining steam generator level with continuous auxiliary 166 246
9 feedwater.
If natural circulation stopped steam pressure would fail quickly as the steam generator cools and steam generator level would rise as long as continuous auxiliary feedwater were supplied.
l(d) Westinahouse Response Although desirable, the use of pressurizer heaters is not the only method for maintaining an overpressure on the reactor coolant system.
The current Westinghouse stanJard configuration since 1970 for insuring pressurization via heater availability is as follows:
During normal operation the pressurizer control or proportional heater group is used automatically to compensate for the reactor coolant system heat losses. The four back-up heater groups are connected to individual on-off co;1 trollers for use during a pressure reduction.
Two back-up heaters groups are required to be, one each, on separate vital power supplies so that no single power failure defeats them.
These vital power supplies (safeguards buses) normally receive power from an outside source.
In case this source fails, power is available frcm the emergency diesel generators.
The energy of the steam and water contained in the pressurizer assures that pressure will be maintaincd to prevent bulk boiling in the core for i sufficient period of time following reactor trip to permit restoration of power to the heaters.
In case the normal power source is lost, the emergency power from the diesel generators is available for the backup heaters which is sufficient to overcome pressurizer heat losses, and keep the RCS pressurized.
1(e) Westinghouse Response Westinghouse concurs that the operator should make use of the steam tables, or preferably, a pressure-temperature chart in the control board.
See the following response for a discussion on the fuel exit thermccouples.
166 247
Recommendation 2 -
Thermocouples used to measure fuel assembly exit temperatures to determine core performance should be used, where currently available, to guide oper-ator concerning core status (full range capabil-ity).
Westinahouse Response The Westinghouse thermocouple system is primarily a surveillance sys-tem. Chromel-alumel thermocouples are threaded into guide tubes that penetrate the reactor vessel head through seal assemblies, and termi-nate at the exit flow end of the fuel assemblies.
All of the control for the thermocouple system is located in the control room.
There are 50 to 65 thermocouples in the system depending on the plant vintage.
There are two direct means of monitoring thermocouple outputs.
- 1) A multipoint precision indicator is provided to indicate tne temperature sensed by the thermocouples.
Only one thermocouple at a time can De connected to the indicator.
- 2) The primary readout is through the plant computer.
The plant ccmputer can accept all of the thermocouple outputs simultaneously and can print out the information upon demano.
The current design for both the indicator and computer outputs limits 0
the readout of the thermocouple system to 750 F.
The thermocouples themselves are capable of measuring temperature to 0
about 2300 F, although the accuracy may be impaired.
However, the connectors, reference junction box, or sneathing may be more limiting than th.e thermocouples.
Information is being gathered to cetermine the limiting component.
A milivolt meter can be used as an alternative to take direct readings at the reference junction box and convert the results into temperatures using the calibration tables that are readily available.
166 248
Recommendation 3 -
Operating reactors should be given priority regard-ing definition and implementation of instrumenta-tion to diagnose and follow the course of a serious accident, includii.g (a) improved sampling procedures under accident conditions (b) improved techniques to provide guidance to offsite authorities.
Westinghouse Response Westinghouse does not normally provide the systems, components or pro-cedures for sampling, and is not involved in the generation of the Emer-gency Plans. Westinghouse does not feel it is appropriate to comment on these ACRS recommendations.
Westinghouse is evaluating information available to the operator to diagnose the event, confirm diagnose, and verify system response. This evaluation will provide instrumer.t qualification requirements, ranges, and accuracies.
166 249
Recommendation 4 -
Reiterates previous reccmmendations that high priority be given to "research to improve reactor safety" (a) research on behavior of LWR's during ancmalous transients (b) NRC to develop capability to simulate wide range of postulated transients and accident conditions.
Westinahouse Response Westinghouse does support and is actively involveu in industry-wide safety research on the behavior of light water reactors.
We participate in many NRC and EPRI sponsored programs.
For example, we have considec-able resources committed to the FLECHT-SEASET and Blowdown Heat Transfer programs. We also closely follow industry development of neu analytical models, such as advanced best estimate analyses like TRAC, and of tests sucn as those at the LOFT facility. Westinghouse is also fully commit-ted to an extensive joint R&D program with Framatcme, CEA, and EDF in France.
Recent programs there have investigated two-phase flow pump characteristics and steam / water mixing tests.
Westinghouse also conducts theoretical analyses, exploration, experi-mentation, and verification testing to demonstrate the safety or our designs and to more sharply define margins of conservatism ano to improve designs.
The Committee is well aware of the various fuel assem-bly tests conducted by Westinghouse and reported on at recent ARCS meet-ings. We have also emoarked on a major prototype test program for our new instrumentation and control system, such as the Integrated Protec-tion System, prior to their introduction into a nuclear power plant.
Westinghouse continues to place a high priccity on participation in industry-wice research, on Corporate experimental and analytical devel-opment, and on verification testing to improve reactor safety.
166 250
Reccmmendation S -
Consideration should be given to additional moni-toring of ESF equipment status, and to supporting services, to help assure availability at all times.
Westinghouse Response We agree that the operator needs at all times assurance of the available of various engineered safety features (ESF) and their supporting systems.
In view of the TMI experience, Westinghouse is reviewing tne matter of status monitoring of ESF systems to assure availability and is consider-ing whether additional recommendations or actions are indicated in this regard.
166 251
C.
Letter, M. Carbon to Acting Chairman Gilinsky dated Aoril 20, 1979 Recommendation 1 -
Initiate immediately a survey of operating proce-dures (p achieving natural circulation, including:
(a) event involving loss of offsice power (b) consideration of role of PZR heaters.
West.1ghouse Respcnse See the previous response to recommendation 1 from the ACRS letter dated April 18, 1979.
T66 252
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